AP1000蒸发器传热管破裂事故分析及敏感性研究

发布时间：2019-02-11 14:40

【摘要】：本文以AP1000核电厂为对象,首先利用系统分析程序RELAP5进行建模,参考西屋公司的SGTR事故进程,设定了安全系统以及辅助系统的促发逻辑和顺序,对AP1000 SGTR事故进行模拟计算,将稳态与瞬态(单根传热管破裂)计算结果和西屋公司SGTR事故分析结果进行比对分析;然后在单根传热管破裂事故模型的基础上,进一步对SGTR事故进行深入分析,研究了多根传热管破裂事故,为验证电厂非能动安全特性,考察了厂外电是否有效和完好侧大气释放阀是否开启故障对事故后果的影响;最后针对破口模型和传热管节点划分数量进行了敏感性分析。结果表明,本文基于RELAP5建立的模型和西屋公司LOFTTR2的计算结果具有较好的一致性,稳态热工参数比如压力、温度、流量等都吻合良好,在单根传热管破裂事故工况下,瞬态参数的变化趋势大致相同,由于物理模型的不同,在数值上存在一定的差异,AP1000依靠非能动余热排出系统能够避免破损SG发生满溢,且有一定的裕量,多根传热管破裂工况下,堆内有可能会出现两相从而导致流动不稳定的现象出现,需要加以关注,改变假设条件进行特性分析及敏感性研究对事故后果都有不同程度的影响,采用不同的破口模型会使事故进程及冷却剂丧失质量发生改变,节点划分数量的不同影响破口临界喷放流量的大小,但系统响应大致相同,破损侧SG均没有发生满溢,更加验证了三代核电技术的安全性,其研究结果可进一步支持AP1000相关的审评工作。
[Abstract]:In this paper, AP1000 nuclear power plant is taken as the object. Firstly, the system analysis program RELAP5 is used to model the model. Referring to the SGTR accident process of Westinghouse Company, the trigger logic and sequence of safety system and auxiliary system are set up, and the simulation calculation of AP1000 SGTR accident is carried out. The calculation results of steady state and transient state (single heat transfer tube rupture) and SGTR accident analysis results of Westinghouse Company are compared and analyzed. Then on the basis of the single heat transfer tube rupture accident model, the SGTR accident is further analyzed, and the multiple heat transfer tube rupture accident is studied, in order to verify the non-active safety characteristics of the power plant. The effect on the accident consequence of whether the power plant is effective or not and whether the air release valve of the intact side is open or not is investigated. Finally, sensitivity analysis is carried out for the break model and the number of heat transfer pipe nodes. The results show that the model based on RELAP5 is in good agreement with the calculation results of LOFTTR2 of Westinghouse, and the steady-state thermal parameters such as pressure, temperature and flow rate are in good agreement. The variation trend of transient parameters is roughly the same. Because of the difference of physical model, there are some differences in numerical value. AP1000 can avoid the overflowing of damaged SG and have a certain margin by relying on inactive residual heat discharge system. Under the condition of rupture of multiple heat transfer pipes, it is possible that two phases will appear in the reactor, which will lead to flow instability, which needs to be paid attention to. Changing the hypothetical conditions for characteristic analysis and sensitivity study have different degrees of influence on the consequence of the accident. Different fracture models will change the process of the accident and the quality of coolant loss, and the number of nodes will affect the critical discharge rate of the break, but the response of the system is roughly the same, and the SG of the damaged side is not overflowing. The safety of the third generation nuclear power technology is verified, and the research results can further support the AP1000 related review.
【学位授予单位】：哈尔滨工程大学
【学位级别】：硕士
【学位授予年份】：2014
【分类号】：TM623.4

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